• 기술혁신학회지
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• 제2권3호
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• pp.118-128
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• 1999

This study attempts to provide an efficient method for managing the Nuclear R&D projects. It should be noted that the technology and experiences to develop a well-balanced plan and mon-itoring for the R&D project are not yet well established when compared with the commercial con-structionand engineering projects. This study recommends that the R&D project manager utilize the scheduling techniques to establish an integral management system taking in to account the un-certainties in the research environment. Based upon the results of a case study for the Liquid Metal Reactor Design Technology Development Project which is led by the KAERI this study also con-firms that it is possible to run the national nuclear R&D projects with the Scheduling & Planning.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2004년도 유체기계 연구개발 발표회 논문집
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• pp.473-478
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• 2004

This work presents the Thermal Hydraulic analysis has been performed for a 19-pin wire-spacer fuel assembly using three-dimensional Reynolds-averaged Navier-Stokes equations. SST model is used as a turbulence closure. The whole fuel assembly has been analyzed for one period of the wire-spacer using periodic boundary condition at inlet and outlet of the calculation domain. The overall results far a preliminary calculation show a good agreement with the experimental observations. It has been found that the major unidirectional flows are the axial velocity in sub-channels and the peripheral sweeping flows and the velocities are found to be following a cyclic path of period equal to the wire-wrap pitch. The temperature is found to be maximum in the central region and also, there exist a radial temperature gradient between the fuel rods. The major advantage of performing this kind of analysis is the prediction of thermal-hydraulic behavior of a fuel assembly with much ease.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1999년도 추계학술대회 논문집 전문대학교육위원 P
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• pp.1-3
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• 1999

EM (ElectroMagnetic) pump is used for the purpose of transporting liquid sodium coolant with electrical conductivity in the LMR(Liquid Metal Reactor). (In the present study, pilot EM pump has been designed by using of equivalent circuit method which is commonly employed to analyze linear induction machines for the test of removal of residual heat. The length and diameter of the pump have fixed values of 840 mm and 101.6 mm each by taking account of geometrical size of circulation loop for the installation of EM pump. Flowrate versus developing pressure is related from Laithwaite's standard design formula and the characteristic analyses of developing force and efficiency are carried out according to change of input frequency. From the characteristic curve, input frequency of 13 Hz is determined as the design frequency. On the other hand, The annular air gap size of 6.05 mm is selected not to bring about too much hydraulic loss. Resultantly design analysis makes pump have the electrical input of 604 VA and the hydrodynamical capacity of 1.3 bars and 40 l/min.

• 에너지공학
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• 제13권3호
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• pp.181-190
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• 2004

Computational models for analyzing the in-reactor behavior of metallic fuel pins under transient conditions in liquid-metal reactors are developed and implemented in the TRAMAC (TRAnsient thermo-Mechanical Analysis Code) for a metal fuel rod under transient operation conditions. Not only the basic models for a fuel rod performance but also some sub-models used for transient condition are installed in TRAMAC. Among the models, a fission gas release model, which takes the multi-bubble size distribution into account to characterize the lenticular bubble shape and the saturation condition on the grain boundary and the cladding deformation model have been developed based mainly on the existing models in the MAC-SIS code. Finally, cladding strains are calculated from the amount of thermal creep, irradiation creep, and irradiation swelling. The cladding strain model in TRAMAC predicts well the absolute magnitudes and gen-eral trends of their predictions compared with those of experimental data. TRAMAC results for the FH-1,2,6 pins are more conservative than experimental data and relatively reasonable than those of FPIN2 code. From the calculation results of TRAMAC, it is apparent that the code is capable of predicting fission gas release, and cladding deformation for LMR metal fuel finder transient operation conditions. The results show that in general, the predictions of TRAMAC agree well with the available irradiation data.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1998년도 추계학술대회 논문집 학회본부A
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• pp.13-15
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• 1998

EM(Electro Magnetic) pump is used for the purpose of transporting liquid sodium coolant with electrical conductivity in the LMR(Liquid Metal Reactor). In the present study. pilot EM pump has been designed by using of equivalent circuit method which is commonly employed to analyze linear induction machines for the test of removal of residual heat. The length and diameter of the pump have fixed values of 840 mm and 101.6 mm each by taking account of geometrical size of circulation loop for the installation of EM pump. Flowrate versus developing pressure is related from Laithwaite's standard design formula and the characteristic analyses of developing force and efficiency are carried out according to change of input frequency. From the characteristic curve, input frequency of 13 Hz is determined as the design frequency. On the other hand, The annular air gap size of 6.05 mm is selected not to bring about too much hydraulic loss. Resultantly design analysis makes pump have the electrical input of 604 VA and the hydrodynamical capacity of 1.3 bars and 40 l/min.

• 에너지공학
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• 제13권2호
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• pp.101-111
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• 2004

액체금속로 노심 열유체 설계의 기본 목표는 노심을 구성하는 집합체에서 발생하는 열량을 효과적으로 추출하기 위해 각각의 집합체 냉각재 유량을 적절히 분배하고, 이에 따른 온도분포가 적절하게 유지되도록 하는 것이다. 노심 열유체 설계 및 특성 분석은 전체노심에 대한 각 집합체의 유량영역을 구분하고, 집합체별 온도분포를 계산하여, 최종적으로 집합체에 대한 상세 부수로 해석을 하는 과정으로 진행된다. 본 논문에서는 이러한 액체금속로의 노심 열유체 설계 방법론을 기술하고, 이를 바탕으로 KALIMER의 증식특성 노심과 breakeven 노심에 대한 열유체 설계와 특성분석을 수행하였다. KALIMER는 원자력 중장기 과제로 개념설계가 진행 중인 전기출력 150MWe, 열출력 392MWth의 금속핵연료를 사용하는 액체 금속로이다.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1053-1064
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• 2009

The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect.

• 한국지진공학회논문집
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• 제5권5호
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• pp.25-33
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• 2001

현재 국내에서 개발중인 액체금속로 원자로구조물의 내진설계 및 지진해석을 위하여 원자로내부에 존재하는 유체-구조물 상호작용을 고려한 단순지진해석 모델링 개발이 필수적이다. 이를 위하여 본 논문에서는 유체속에 잠긴 동축원통 구조물의 유체부가질량에 대한 이론적 배경을 검토하였으며 기존의 유체부가질량법을 고려하여 시간이력 지진응답해석을 수행할 수 있는 Runge-Kutta 수치해석 알고리즘을 사용한 해석코드를 개발하였다. 개발된 지진해석코드를 사용하여 유체속에 잠긴 두개의 동축원통 구조물의 진동특성과 지진응답에 대한 유체부가질량의 영향을 살펴보았다. 적용예로서 두개의 동축원통에 대한 단순지진해석모델을 가정하고 유체부가질량을 고려한 진동특성해석과 지진응답해석을 수행한 결과 유체가 채워진 동축원통 구조물은 유체와의 상호작용으로 인하여 구조물의 진동특성과 지진응답특성이 크게 영향을 받으며 지진응답해석시 연계항을 포함한 유체부가질량의 영향을 신중하게 고려할 필요가 있는 것으로 나타났다.

• 한국재료학회지
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• 제9권11호
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• pp.1137-1143
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• 1999

액체금속로 전열관재료용 개량 9Cr-1Mo강의 특성에 미치는 텅스텐의 영향을 고찰하기 위하여 개량 9Cr-1Mo강에 텅스텐을 2wt.% 첨가하여 템퍼링 온도에 따른 기계적 특성 및 미세조직의 변화를 조사하였다. 미세조직을 관찰한 결과 템퍼링시 전위회복에 의해 형성되는 셀 구조가 나타나는 템퍼링 온도는 개량 9Cr-1Mo강의 경우 $700^{\circ}C$인 반면, 텅스텐을 첨가한 9Cr-0.5Mo-2W강의 경우는 $750^{\circ}C$이었으며, 이 결과로부터 텅스텐 첨가는 전위회복을 지연하였음을 알 수 있다. 텅스텐을 첨가하여도 템퍼링 온도에 따라 생성되는 석출물의 종류는 차이가 없었으나, 텅스텐을 첨가한 강에 있는 석출물에는 텅스텐이 포함되어 있었다. 텅스텐의 첨가로 경도값, 고온 인장강도 그리고 항복강도가 증가하였다. 이것은 텅스텐 첨가로 인한 미세구조의 안정화에 기인하는 것으로 보인다. 충격시험에서는 항복강도가 낮은 개량 9Cr-1Mo강이 더 우수한 충격파괴특성을 가졌다.